The present invention relates to alloys based on zirconium to be employed for example as construction members in the nuclear reactor of a nuclear power plant etc. and to a method for treating these alloys.
A typical fuel assemply employed in a nuclear power plant has a construction as generally shown in the elevational view of appended FIG. 2 in which a plurality of fuel elements 7, constructed as shown in FIG. 1, in a vertical section are assembled in a form of upright lattice. Nuclear fuels 1 each consisting of a cylindrical sintered product of uranium oxide (denoted hereinafter as pellet) are packed in a sheath tube 2 sealed at both ends with terminal stoppers 4, 5. A coil spring 3 tightens the pellets 1 within the sheath tube 2. A number of the so constructed fuel elements 7 are supported on support gratings 6 and arranged to build up a fuel assembly having an upper nozzle 8, a bottom nozzle 9, a suspending diaphragm spring 10 and a control rod cluster 11.
As the materials for the sheath tube 2 of the fuel element 7 and for the support grating 6, zirconium alloys R60802 or R60804 of UNS (Unified Numbering System for Metals and Alloys) defined by ASTM B353 (designated hereinafter as Zircaloy 2 and Zircaloy 4 for the former and for the latter respectively) have conventionally been employed, wherein the former is a zirconium alloy containing tin, iron, chromium and nickel each in a small amount.
During running of a nuclear power plant, the outer surfaces of internal construction members in the nuclear reactor are held in contact with the cooling water maintained at high temperature under high pressure so that the materials consisting of zirconium alloys constitutung these members will be subjected to corrosion, namely, a high temperature reaction with hot water or with high temperature steam to form a uniform or lacal oxide cover layer while the hydrogen formed thereby penetrates through the oxide layer and is absorbed in the alloy. Upon progression of such reaction (hereinafter denoted as corrosion) with corresponding thickening of the oxide cover layer and decrease in the virtual thickness of the body of alloy, the mechanical strength of construction members, such as sheath tube 2, support grating 6 and so on, made of zirconium alloy, decreases. In addition, the strength and ductility of zirconium alloy construction members decrease with the increase in the amount of hydrogen absorbed in the alloy, which is formed by the above-mentioned corrosion reaction. Thus, the corrosion of sheath tubes 2 and support gratings 6 may result in a reduction of the performances of these members due to the decrease in the strength and the ductility. In the practice however, the extent of corrosion on the outer surfaces of sheath tubes and support gratings is quite small under the running condition of nuclear power plant of nowadays and, thus, has not reached hitherto any failure in the proper functions of these members.
However, there has been, in fact, a problem, as mentioned above, of possible occurrence of failure in the performances of the members made of zirconium alloys, especially in the case where the fission rate of the nuclear fuel is increased and the retention time of the fuel within the nuclear reactor is extended so as to attain an efficient utilization of the nuclear fuel, due to the progression of the above-mentioned corrosion on the outer surfaces of construction members, such as sheath tubes and support gratings made of zirconium alloys, resulting in a thickening of the zirconium oxide outer layer with corresponding decrease in the virtual thickness of the zirconium alloy body, together with the existing danger of destruction of proper functions of the zirconium alloy construction members by the increase in the amount of hydrogen absorbed in the alloy.
In order to improve the corrosion resistance of zirconium alloys. a countermeasure had been proposed, as disclosed in the Japanese patent application Ser. No. Sho 62-46709, in which the amounts of subsidiary alloy elements tin, iron, chromium and niobium are adjusted adequately. It was reported, in particular, that a marked increase in the corrosion resistance was recognized by reducing the content of tin.
By such an adjustment of the contents of subsidiary alloy elements, however, change in the mechanical properties, such as strength etc., of the alloy material will result also. In particular, by reducing the content of tin, a decrease in the creep strength as compared with that of conventional materials of the sheath tube of fuel element and the support grating will result, so that there occurs a danger of greater decrement in the outer diameter of the fuel element to occur during the operation of the nuclear reactor when using such a material for the sheath tube, whereby a possibility occurs that the sheath tube or the like will be loaded by too high a stress upon an abrupt increase in the output power.